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| Quarter | Site | Title | Description | |
|---|---|---|---|---|
| 05000293/FIN-2018003-01 | 2018Q3 | Pilgrim | Failure to Identify an Adverse Condition Associated with Elevated Standby Gas Treatment System Accumulator Leakage | The inspectors identified a Green non-cited violation (NCV) of Technical Specifications 3.7.B.1.c because Entergy exceeded the TS allowed outage time for the standby gas treatment system (SBGT) when the station did not identify an adverse condition associated with elevated air accumulator leakage in the system. |
| 05000293/FIN-2018002-06 | 2018Q2 | Pilgrim | Minor Violation | This violation of minor significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a minor violation, consistent with the NRC Enforcement Policy. On June 22, 2015, Entergy submitted a licensee event report in accordance with 10 CFR 50.73 that contained information that was not complete or accurate in all material respects, contrary to the requirements in 10 CFR 50.9. Specifically, the licensee submitted Licensee Event Report 2015-004-00 to communicate the failure during testing of time delay Agastat relay 27A-B1X/TDDO intended to provide under-voltage protection for 480V emergency bus B6 by transferring power from bus B1 to bus B2. In the licensee event report, Entergy incorrectly documented that due to the failure, bus B6 would have continued to receive power from bus B1 with degraded voltage. Upon identifying the issue, on March 8, 2016, Entergy submitted a revised licensee event report with the correct information. Enforcement: 10 CFR 50.9 requires that information provided to the Commission by a licensee shall be complete and accurate in all material respects. Contrary to the above, on June 22, 2015, Entergy provided information to the Commission that was not complete and accurate in all material respects. In the licensee event report, the licensee documented that due to the failure, bus B6 would have continued to receive power from bus B1 with degraded voltage. However, bus B6 would actually have tripped from bus B1 and lost power completely. This information was material to the NRC because the NRC requires timely and accurate reporting of information related to events in order to evaluate the potential safety significance and required NRC response. Entergy identified the inaccuracy and entered the issue into its corrective action program (CR-PNP-2015-9762). On March 8, 2016, Entergy submitted a revision to the licensee event report (2015-004-01) that corrected the report. This failure to comply with 10 CFR 50.9 constitutes a minor violation that is not subject to enforcement action in accordance with the NRCs Enforcement Policy. The disposition of this violation closes Licensee Event Report 05000293/2015-004-01. |
| 05000293/FIN-2018002-05 | 2018Q2 | Pilgrim | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: 10 CFR 50.72(b)(3)(v)(C) requires licensees to a notify the NRC within 8 hours any event or condition that at the time of discovery could have prevented the fulfillment of the safety function of structures or systems that are needed to control the release of radioactive material. Contrary to the above, Entergy did not make a required notification pursuant to 10 CFR 50.72(b)(3)(v)(C). Specifically, on June 20, 2017, secondary containment was declared inoperable due to simultaneous opening of both airlock doors, and Entergy did not make the required notification until June 22, 2017. Significance/Severity: This violation is being treated under the NRCs traditional enforcement process, for impeding the regulatory process, specifically Entergy did not make a required notification, as outlined in Inspection Manual Chapter 0612, Appendix B. The Reactor Oversight Processs significance determination process does not specifically consider the regulatory process impact in its assessment of licensee performance. Therefore, it is necessary to address this violation which impedes the NRCs ability to regulate using traditional enforcement to adequately deter non-compliance. The severity of this violation was determined to be Severity Level IV, as outlined in Example 9 from Section 6.9.d. of the NRC Enforcement Policy. Corrective Action References: CR-PNP-2017-06380 and CR-PNP-2017-07015 The disposition of this finding closes Licensee Event Report 2017-011-00. |
| 05000293/FIN-2018002-04 | 2018Q2 | Pilgrim | Licensee-Identified Violation | This violation of very low safety significance was identified by the licensee and has been entered into the licensees corrective action program and is being treated as a NCV, consistent with Section 2.3.2 of the Enforcement Policy. Violation: 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires, in part, that activities affecting quality shall be prescribed by documented instructions appropriate to the circumstances and shall be accomplished in accordance with the instructions. Contrary to the above, from January 1994 to June 2017, Entergy modified site surveillance procedure 8.M.3-18, Standby Gas Treatment System Exhaust Fan Logic Test and Instrument Calibration, without prescribing adequate documented instructions for the condition caused by the testing. Specifically, Entergy failed to identify that the procedurally prescribed lineup of the standby gas treatment system resulted in secondary containment being inoperable due to the large opening introduced into the system. Significance/Severity: The inspectors evaluated this finding using Inspection Manual Chapter 0609.04, Initial Characterization of Findings, and Inspection Manual Chapter 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined that the finding was of very low safety significance. Corrective Action Reference: CR-PNP-2017-11714 The disposition of this violation closes Licensee Event Reports 05000293/2017-013-00 and 05000293/2017-013-01. |
| 05000293/FIN-2018002-03 | 2018Q2 | Pilgrim | 480V Bus B6 Auto Transfer Function Degraded Due to Time Delay Relay Failure | The inspectors identified a Severity Level IV NCV of TS 3.5.A.2 because a component of the low pressure coolant injection system was inoperable between May 12, 2015, and May 3, 2017, during which time, on occasions, core spray systems were also not operable. Specifically, a relay, used to transfer the power feed for the low pressure coolant injection valves to the backup source in the event of a degraded voltage condition, failed during testing. As a result, under certain conditions, the transfer would not have automatically occurred. This condition existed through the operating cycle, during which time the core spray pumps were also inoperable when removed from service for scheduled maintenance. |
| 05000293/FIN-2018002-02 | 2018Q2 | Pilgrim | Loss of Secondary Containment Integrity due to Simultaneously Opened Airlock Doors | A self-revealed Green finding was identified when personnel did not implement a procedure requiring the closure and verification of doors credited with specific design functions. Procedure 1.3.135, Control of Doors, requires station personnel to ensure closing and latching of doors. Failure to meet this requirement caused the loss of secondary containment integrity and unplanned entry into Technical Specification (TS) condition 3.7.C.1. |
| 05000293/FIN-2018002-01 | 2018Q2 | Pilgrim | Failure to Properly Implement the Fatigue Management Program Work Hour Controls for Covered Workers | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 26.205(d). During the period December 2017 to April 2018, Entergy did not properly control the work hours of several workers who performed work covered under 10 CFR 26.4(a). Specifically, on eleven occasions, workers exceeded one of the following work hour limits: (1) 16 work hours in any 24-hour period; (2) 72 hours in any 7-day period; or (3) 54 hours per week average over a 6-week rolling time period. |
| 05000317/FIN-2017004-01 | 2017Q4 | Calvert Cliffs | Inadequate Assessment of Fire Brigade Performance During an Announced Fire Drill | An NRC-identified Green non-cited violation (NCV) of Calvert Cliffs Nuclear Power Plant Renewed Facility Operating License DPR-53, DRP-69, Condition E, was identified for Exelons failure to adequately assess the performance of the fire brigade during an announced fire drill. Specifically, Exelon failed to properly assess the command and control performance of the fire brigade leader (FBL) which resulted in the fire drill being improperly evaluated as having met the assessment criteria. The inspectors determined that Exelons failure to properly assess fire brigade performance in accordance with OP-AA-201-003, Fire Drill Performance, Revision 16, was a performance deficiency. Exelon has entered this issue into their corrective action program (CAP) as action request (AR) 04094397The inspectors reviewed IMC 0612, Appendix B, Issue Screening, issued on September 7, 2012, and determined the issue is more than minor because it was associated with the protection against external events (fire) attribute of the Mitigating Systems cornerstone and adversely affected its objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, failure to properly evaluate the performance of the fire brigade and correct identified deficiencies adversely affects the fire brigades ability to protect against the effects of a fire. In accordance with IMC 0609, Attachment 4, Initial Characterization of Findings, issued on October 7, 2016, and IMC 0609, Appendix A, The Significance Determination Process for Findings at Power issued on June 19, 2012, the inspectors determined that this finding is of very low safety significance (Green) since it involved fire brigade training requirements, the fire brigade demonstrated the ability to meet the required times for fire extinguishment for the fire drill scenario, and the finding did not significantly affect the fire brigades ability to respond to a fire. The inspectors determined that the finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Self-Assessment, because Exelon did not conduct a self-critical and objective assessment of the fire brigades performance. Specifically, Exelon failed to conduct a self-critical and objective assessment of the FBLs performance during the fire drill described above. |
| 05000293/FIN-2017002-02 | 2017Q2 | Pilgrim | Reporting of Unplanned Scrams with Complications Performance Indicator for Feedwater Regulating Valve Scram | The inspectors identified an unresolved item (URI) associated with Entergys reporting of Unplanned Scrams with Complications PI data for the third quarter of 2016. Description. On September 6, 2016, PNPS operators initiated a manual reactor scram based on oscillating feed flow as a result of a malfunction with feedwater regulating valve (FRV) A. As a result of high reactor vessel water level, all of the reactor feed pumps tripped, the HPCI and RCIC systems isolated, and a Group 1 isolation signal was present, initiating closure of the MSIVs. In order to maintain pressure control of the reactor, SRV 3B was manually cycled. This event was reported under Licensee Event Report (LER) 05000293/2016-007-00. During the scram response, PNPS operators were required to use an SRV to maintain reactor pressure control, but Entergys submittal of PI data for the third quarter of 2016 does not count the scram as an Unplanned Scram with Complications, which is required by EN-LI-114, Regulatory Performance Indicator Process. This URI is being opened to determine if a performance deficiency exists pending resolution of the differing interpretation of guidance contained in NEI Document 99-02, Regulatory Assessment Performance Indicator Guidance, Revision 7, at the next scheduled Reactor Oversight Process Working Group Meeting. (URI 05000293/2017002-02, Reporting of Unplanned Scrams with Complications Performance Indicator for Feedwater Regulating Valve Scram) |
| 05000293/FIN-2017002-07 | 2017Q2 | Pilgrim | Untimely 10 CFR 50.72 Notification of a Secondary Containment System Functional Failure | An NRC-identified SL IV NCV of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, was identified because both trains of the SBGTS were made inoperable during surveillance testing, and the condition was not reported to the NRC within eight hours of the occurrence, as required by 10 CFR 50.72(b)(3)(v), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. Specifically, on April 5, 2017, while performing TS SR 4.7.C, trains A and B of the SBGTS were made inoperable leading to the inoperability of the Secondary Containment System (SCS). As a corrective action, Entergy personnel performed a causal evaluation. This issue was entered into the CAP as CR 2017-7446. The inspectors evaluated this performance deficiency in accordance with the traditional enforcement process because the issue impacted the regulatory process, in that a condition that could have prevented a safety function was not reported to the NRC within the required timeframe, thereby delaying the NRCs opportunity to review the matter. Using Example 6.9.d.9 from the NRC Enforcement Policy (the failure of a licensee to make a report as required by 10 CFR 50.72 or 10 CFR 50.73), the inspectors determined that the violation was a SL IV violation. Because this violation involves the traditional enforcement process and does not have an underlying technical violation, inspectors did not assign a cross-cutting aspect, in accordance with IMC 0612, Appendix B. |
| 05000293/FIN-2017002-01 | 2017Q2 | Pilgrim | Failure to Follow Procedure Requirements for the Control of a Flood Protection Barrier | An NRC-identified Green finding was identified because Entergy personnel did not follow Procedure 1.3.135, Control of Doors, to adequately control a condenser bay flood protection door. Specifically, on May 22, 2017, Entergy personnel failed to control door 25A, which is designed to mitigate condenser bay flooding to preclude adversely impacting the important to safety instrument air system. Entergys short-term corrective actions included closing the door and providing additional operator training. This issue was entered into the CAP as CR 2017-5746. The performance deficiency is more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. The finding was evaluated using IMC 0609, Appendix A, Exhibit 4, External Events Screening Questions, issued June 19, 2012, with respect to the degraded safety function of the flood barrier door. The finding was determined to be of very low safety significance (Green) because the failure of the flood door was determined to not degrade the instrument air system ability to support the feedwater injection function or the alternate injection through the control rod drive system. This is because the backup diesel driven compressor was available to be started locally and supply the instrument air headers. The finding also did not involve the total loss of any safety function. The finding has a cross-cutting aspect in the area of Human Performance - Procedure Adherence, because Entergy personnel did not follow processes, procedures, and work instructions. Specifically, Entergy personnel did not follow procedural requirements to adequately control flood protection door 25A. (H.8) |
| 05000293/FIN-2017002-03 | 2017Q2 | Pilgrim | Inaccurate Suppression Pool Water Level Instrument not Identified during Post-event Prompt Investigation | An NRC-identified Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, was identified because Entergy staff did not identify and correct a condition adverse to quality related to suppression pool water level indication when the A suppression pool wide range instrument provided inaccurate level indication during the inadvertent suppression pool water level increase event on March 31, 2017. As corrective actions, Entergy entered Technical Specification (TS) 3.2.F, Protective Instrumentation - Surveillance Information Readouts, and repaired the instrument. This issue was entered into Entergys corrective action program (CAP) as condition report (CR) 2017-2965. The performance deficiency is more than minor because if left uncorrected, the performance deficiency had the potential to lead to a more significant safety concern. Specifically, inaccurate level indication during off-normal changing level conditions in the suppression pool could result in operator actions not warranted by plant conditions. The finding is also associated with the Initiating Events cornerstone. Using IMC 0609, Appendix A, Exhibit 1, issued June 19, 2012, The Significance Determination Process for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not cause a reactor trip and a loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution - Identification, because the Entergy organization did not demonstrate an appropriately low threshold for entering problems into their CAP. Specifically, Entergys prompt investigation of the inadvertent suppression pool level increase event did not identify that the A suppression pool wide range level instrument was not indicating properly and required corrective maintenance. (P.1) |
| 05000293/FIN-2017002-04 | 2017Q2 | Pilgrim | Improper System Restoration Results in Suppression Pool Inoperability | A self-revealing Green NCV of TS 5.4.1.a, Procedures, was identified on March 31, 2017, when operators did not follow procedures and caused an inadvertent increase in the suppression pool water level. The inspectors determined that the operators did not restore the core spray system valve line-up as prescribed in Attachment 11 of Entergy Procedure 2.2.20, Core Spray, and the maintenance safety tag clearance sheet. Operator implementation of these documents is directed by Entergy Procedure EN-OP-102, Protective Caution Tagging, section 5.19(4)(b). As corrective actions, Entergy performed additional management oversight of control room operations and performed a root cause evaluation (RCE). This issue was entered into the CAP as CR-2017-2785. The performance deficiency is more than minor because it is associated with the equipment reliability attribute of the Mitigating Systems cornerstone objective and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the out of specification conditions on March 31, 2017, impacted suppression pool reliability because the suppression pool was not maintained within parameters required to ensure operability. Additionally, significant analysis was necessary to show the suppression pool and associated supports remained functional when TS requirements were not met. Using IMC 0609, Appendix A, Exhibit 2, issued June 19, 2012, The Significance Determination Process for Findings At-Power, the inspectors determined the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating structure, system, or component (SSC), the finding did not represent a loss of system and/or function, the finding did not represent an actual loss of a function of a single train for greater than the TS allowed outage time (AOT), and the finding did not represent an actual loss of a function of one or more non-TS trains of equipment. Specifically, the suppression pool, including downcomers and supports, remained functional following the influx of water. The finding has a cross-cutting aspect in the area of Human Performance - Procedure Adherence, because Entergy personnel did not follow processes, procedures, and work instructions. Specifically, Entergy personnel did not follow procedures and work instructions during the restoration of the core spray system. (H.8) |
| 05000293/FIN-2017002-05 | 2017Q2 | Pilgrim | Damper Failure Causes Loss of Secondary Containment | A self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, and TS 3.7.C.2, Containment Systems Secondary Containment, was identified because Entergy did not establish an appropriate interval to overhaul the secondary containment isolation dampers. As a result, the refueling floor supply isolation dampers were operated beyond the recommended overhaul interval and subsequently failed. Entergys corrective actions included cleaning, lubricating, and post-work testing the failed refueling floor supply isolation dampers. This issue was entered into the CAP as CR 2017-0494. The performance deficiency is more than minor because it is associated with the SSC and barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. Specifically, Entergys preventative maintenance (PM) for the refueling floor supply isolation dampers was inadequate to ensure the availability and reliability of SSCs required to maintain secondary containment operable. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the performance deficiency only represented a degradation of the radiological barrier function provided by the reactor building and standby gas treatment system (SBGTS). The finding has a cross-cutting aspect in the area of Problem Identification and Resolution - Resolution, in that Entergy personnel did not take effective corrective actions to address issues in a timely manner. Specifically, in 2016, Entergy personnel identified there were deficiencies in the PM program with technical justifications for deferring PMs. Entergy reasonably had the opportunity to identify which PMs were not performed within recommended guidelines and make appropriate changes as needed. (P.3) |
| 05000293/FIN-2017002-06 | 2017Q2 | Pilgrim | Secondary Containment Testing not performed per Technical Specifications | An NRC-identified Green NCV of 10 CFR Part 50, Appendix B, Criterion XI, Test Control, and TS 4.7.C, Containment Systems Secondary Containment, was identified when Entergy performed a surveillance test requiring a refueling outage while online. Specifically, Entergy performed Procedure 8.7.3, Secondary Containment Leak Rate Test, TS Surveillance Requirement (SR) 4.7.C from February 27, 1997, to April 5, 2017. As corrective actions, Entergy re-performed the test during the April 2017 refueling outage prior to refueling. This issue was entered into the CAP as CR 2017-2900. The performance deficiency is more than minor because it is associated with the configuration control attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protects the public from radionuclide releases caused by accidents or events. Specifically, Entergy intentionally removed the safety function of standby gas and secondary containment for operational convenience and did not comply with the requirements of TS SR 4.7.C which requires the test to be performed during a refueling outage before refueling. In accordance with IMC 0609.04, Initial Characterization of Findings, issued October 7, 2016, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green), because the finding only represented a degradation of the radiological barrier function provided for the SBGTS. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance - Conservative Bias, in that Entergy personnel did not use decision making-practices that emphasize prudent choices over those that are simply allowable. Specifically, operators did not refer to the TSs to understand the required conditions for a secondary containment surveillance test. Operators followed an inadequate site procedure for the plant conditions at the time and did not question why removal of a safety function for operational convenience was acceptable. (H.14) |
| 05000293/FIN-2017002-08 | 2017Q2 | Pilgrim | Licensee-Identified Violation | 10 CFR Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, requires in part, that activities affecting quality shall be accomplished in accordance with documented procedures. Entergy Procedure EN-OP-104, Operability Determination Process, requires that operators have a reasonable expectation of operability when determining the operability of a component. On April 15, 2017, operators did not have a reasonable expectation of operability, as required by EN-OP-104, and incorrectly declared the B SRM operable without reasonable assurance. This resulted in a violation of TS 3.10.B, Core Alterations, which requires, during core alterations, when fuel is in the vessel, at least 2 SRMs shall be operable, one in the quadrant where fuel or control rods are being moved and one in an adjacent quadrant. Entergy entered this issue into the CAP as CRs 2017-3541, 2017-3952, 2017-5294, and 2017-6724. Entergy repaired the B SRM, and performed a causal evaluation on the equipment failure that includes the late inoperability determination by the operators. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix G, Attachment 1, Exhibit 3, Mitigating Systems Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a system, and did not represent a loss of safety function of a train or system, and did not degrade a functional auto-isolation of RHR on low reactor vessel level. |
| 05000293/FIN-2017001-01 | 2017Q1 | Pilgrim | Concern Regarding Ability to Declare EALs during Loss of Control Room Air Conditioning | Inspection Scope The inspectors reviewed the samples listed below to assess the effectiveness of maintenance activities on structure, system, and component (SSC) performance and reliability. The inspectors reviewed system health reports, CAP documents, maintenance WOs, and maintenance rule (MR) basis documents to ensure that Entergy was identifying and properly evaluating performance problems within the scope of the MR. For each sample selected, the inspectors verified that the SSC was properly scoped into the MR in accordance with 10 CFR 50.65 and verified that the (a)(2) performance criteria established by Entergy staff was reasonable. As applicable, for SSCs classified as (a)(1), the inspectors assessed the adequacy of goals and corrective actions to return these SSCs to (a)(2). Additionally, the inspectors ensured that Entergy staff was identifying and addressing common cause failures that occurred within and across MR system boundaries. HPCI stop valve grease on February 17, 2017 (quality control) Main control room ventilation the week of March 6, 2017 9 b. Findings Introduction. The inspectors identified that Entergy made alterations on February 2, 2017, to procedure 2.4.149, Loss of Control Room Air Conditioning, that had the potential to render several emergency action levels (EALs) ineffective. As a result, the NRC opened an unresolved item related to this concern. Description. The inspectors identified a concern regarding Entergys ability to declare several EALs based on the actions required by site procedure 2.4.149, Loss of Control Room Air Conditioning. Specifically, procedure 2.4.149 directs numerous loads to be shed in order to maintain the main control room temperature below 120 degrees Fahrenheit upon loss of control room air conditioning during extended period of outside temperature of 90 degrees Fahrenheit and above, as per FSAR section 7.1.8. Main control room air conditioning is not consider ed important to safety, based on the ability to control the heat up rate in the main control room, through the actions described in procedure 2.4.149. Upon updating the calculation to determine how much load must be shed to ensure design requirements were met, procedure 2.4.149 was updated with an attachment directing which loads that are required to be shed in order to meet the design calculation S&SA056, Control Room and Cable Spreading Room Heatup Calculations, Revision 6. The main control room is required to remain at or below 120 degrees Fahrenheit to ensure the main control room equipment remains operable. Main control room equipment temperatures above 120 degrees Fahrenheit can result in multiple control equipment failures which could result in misleading indications and inadvertent system actuation. The inspectors questioned how the procedure would be implemented, based on the lack of specific guidance in the procedure. The procedure includes the load shedding of numerous components, including both trains of reactor protection system, average power range monitors, intermediate range power monitors, source range power monitors, and process radiation monitors. Inspectors questioned how the site would declare numerous EALs without supporting equipment that has no redundancy or pre- established compensatory measures, as proceduralized in EN-AD-270, Equipment Important to Emergency Response. Inspectors questioned at what point would the operators be required to shed equipment that is required to support the HOT (greater than 212 degrees Fahrenheit) condition EAL classifications. The inspectors questioned whether or not operators would be able to verify that the plant conditions were consistent with applicable EALs at the time the components were removed from service. Entergy is reviewing the calculations to determine when load shedding of loads without compensatory measures would have been required and intends to report the results to the NRC by June 2, 2017. Inspectors verified that the procedure was changed to ensure minimum instrumentation requirements were maintained to declare EALs. The inspectors determined that procedure 2.4.149 had the potential to render EALs ineffective and is an unresolved item pending Entergy completing their evaluation of load shedding impact on the main control room heat up and NRC review of the evaluation and procedure implementation. (URI 05000293/2017001-01, Concern Regarding Ability to Declare EALs during Loss of Control Room Air Conditioning) |
| 05000293/FIN-2017001-04 | 2017Q1 | Pilgrim | Failure to Submit a Required 50.72 Notification | Severity Level lV. The inspectors identified a Severity Level IV NCV of 10 CFR 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, because a TS required shutdown was not reported to the NRC within four hours of the occurrence, as required by 10 CFR 50.72(b)(2)(i). Specifically, on December 16, 2016, PNPS initiated a shutdown, as required by TS, as a result of the discovery of leakage associated with main steam isolation valves (MSIVs) 2C and 2D, leadi ng to the required isolation of the C and D main steam lines. Entergy entered the issue into the CAP as CR 2017-3723. Inspectors determined the issue had the potential to affect the NRCs ability to perform its regulatory function, therefore, the inspectors evaluated this performance deficiency in accordance with the traditional enforcement process. Using example 6.9.d.9 from the NRC Enforcement Policy (the failure of a licensee to make a report as required by 10 CFR 50.72 or 10 CFR 50.73), the inspectors determined that the violation was a Severity Level IV violation. Because this violation involves the traditional enforcement process and does not have an underlying technical violation, inspectors did not assign a cross-cutting aspect to this violation in accordance with IMC 0612, Appendix B |
| 05000293/FIN-2017001-03 | 2017Q1 | Pilgrim | Untimely Corrective Actions associated with Boraflex degradation in the Spent Fuel Pool | Green. The inspectors identified a Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Actions, when Entergy did not take timely corrective action to correct a condition adverse to quality. Specifically, when BADGER testing results revealed gaps in 4 neutron absorber material that exceeded spent fuel storage design feature assumptions and therefore did not ensure compliance with TSs, the station did not establish corrective actions to ensure configurations and limitations would meet subcriticality analysis beyond September 2017. Entergy entered this into the CAP as CR 2017-1650 and is performing a root cause evaluation to evaluate options and establish corrective actions to ensure compliance is met beyond this timeframe. The performance deficiency was more than minor because it was associated with the Barrier Integrity cornerstone attribute of configuration control (reactivity control) and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding is of very low safety significance (Green) because the finding did not adversely affect any of the barrier integrity screening questions. The inspectors determined this finding had a cross-cutting aspect in Problem Identification and Resolution, Evaluation, because the organization did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the casual evaluation written to address the boraflex degradation was focused on restoring compliance and correcting immediate condition, and did not include longer term corrective actions to mitigate the likelihood of recurrence. (P.2) |
| 05000293/FIN-2017001-02 | 2017Q1 | Pilgrim | Failure to Follow Procedures for Controlled Shutdown | Green. The inspectors identified a Green NCV of TS 5.4.1 Procedures, when Entergy did not follow the site procedures for limiting condition for operation (LCO) entries, Technical Specification (TS) usage, and procedure adherence. Specifically, on March 1, 2017, Entergy did not implement procedure 1.3.6, Technical Specifications-Adherence and Clarifications, and perform the procedural required preparation steps to commence a controlled and orderly shutdown when required by TS LCOs. Additionally, Entergy did not properly exit a TS LCO, based on procedure 1. 3.34.2, Limiting Conditions for Operation Log, requirements. Entergy entered the issue into the corrective action program (CAP) as condition report (CR) 2017-3724. The performance deficiency is more than minor because if left uncorrected, would have the potential to lead to a more significant safety concern. Specifically, the Entergy operations staff exited the LPCI LCO without personal observation by the senior reactor operator (SRO) signing off the work order (WO) that the maintenance postwork testing was complete and failed to implement the procedural required preparation steps to perform a controlled and orderly shutdown when required by TS LCOs. Inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, and determined that the finding was of very low safety significance (Green), because the finding was not a design or qualification deficiency, did not represent a loss of safety system function, and did not screen as potentially risk significant due to external initiating events. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, in that individuals follow processes, procedures, and work instructions. Specifically, Entergy did not us e procedural guidance explicitly put in place to provide operators clear direction on how to prepare and perform an orderly shutdown upon entering a TS LCO with shutdown requirements. (H.8) |
| 05000293/FIN-2016004-02 | 2016Q4 | Pilgrim | Ineffective Corrective Actions to Correct High Pressure Coolant Injection System Vibrations | Green. A self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified in that Entergy did not identify and correct a condition adverse to quality related to high pressure coolant injection (HPCI) pump degraded performance, as required by EN-LI-102, Corrective Action Program. EN-LI-102, requires, in part, that individuals closing corrective actions verify that the required action has been taken ensuring that the response is adequate, answers all aspects of the assigned action, and the intent of the action is met. Specifically, vibrations on the HPCI main pump to speed reducer coupling were not addressed during HPCI system maintenance, despite a degrading trend starting May 21, 2015. This led to the HPCI system being declared inoperable on November 7, 2016, after vibration levels exceeded the in-service testing (IST) action range threshold. Entergys corrective actions included modeling vibrations of the HPCI system during operation and installing a stiffening plate on the HPCI pump support pedestal in order to dampen vibrations associated with the system. Entergy has entered this into their CAP as CR 2016-8657. The inspectors determined that this performance deficiency was more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and adversely affected its objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage.) Specifically, Entergy did not address the increase in HPCI pump vibrations from May 21, 2015, to November 7, 2016, when the vibrations increased into the IST Action range and resulted in pump inoperability. In accordance with IMC 0609.04, Initial Characterization of Findings, issued October 7, 2016, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green) because the finding did not affect the design or qualification of a mitigating structure, system, or component (SSC), represent a loss of system and/or function, involve an actual loss of a function of at least a single train or two separate safety systems for a greater time than allowed by technical specifications (TS), or represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with the licensees maintenance rule program. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Design Margins, in that the organization operates and maintains equipment within design margins, and margins are carefully guarded and changed only through a systematic and rigorous process. Specifically, Entergy did not demonstrate that the work process supports nuclear safety and maintenance of design margins by minimizing long-standing equipment issues, preventive maintenance (PM) deferrals, and maintenance and engineering backlogs. Entergys failure to effectively manage design margins regarding HPCI system vibrations led to a continuing degradation of the system, and the eventual need to declare the HPCI system inoperable on November 7, 2016. (H.6) |
| 05000293/FIN-2016004-01 | 2016Q4 | Pilgrim | Failure to Promptly Perform an Operability Evaluation for a Recirculation Flow Converter | Green. The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergy did not perform a prompt operability determination and adequately evaluate the operability of a recirculation flow converter in a timely manner in accordance with procedure EN-OP-104, Operability Determination Process. As a result, Entergy allowed this flow converter to remain in service, without reasonable assurance of its capability to perform its required safety function, from the time the adverse condition was discovered on October 3, 2016, until the component was declared inoperable and replaced on October 21, 2016. Entergy entered the initial equipment failure into the CAP as CR 2016-07622 and CR 2017-0854. Entergy took corrective actions to replace the inoperable flow converter. The inspectors determined that this performance deficiency was more than minor because it was associated with the equipment performance attribute of the Mitigating Systems cornerstone and it adversely affected the cornerstone objective of ensuring the reliability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). The issue is also similar to the more than minor example in IMC 0612, Appendix E, Examples of Minor Issues, issued August 11, 2009, Example 3j because the flow converters capability to perform its required safety function could not be reasonably assured. The inspectors screened this finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At- Power, issued June 19, 2012, Exhibit 2, Mitigating Systems Screening Questions, and determined that this finding was of very low safety significance (Green) because the finding affected a single reactor protection system (RPS) trip signal to initiate a reactor scram, but did not affect the function of other redundant trips or diverse methods of reactor shutdown, did not involve control manipulations that unintentionally added positive reactivity, and did not result in a mismanagement of reactivity by operator. The inspectors determined that this finding had a -cutting aspect in the area of Human Performance, Conservative Bias, because Entergy did not use decision makingpractices that emphasize prudent choices over those that are simply allowable. Specifically, Entergy did not take a conservative approach in making the decision to keep the A recirculation flow converter in service when available information regarding its operability was incomplete. Operators continued to act based on the assumption that the flow converter would remain operable, without reasonable assurance. Management did not adequately prioritize the completion of the operability evaluation for this safetyrelated component. Instead, the completion of the evaluation was delayed due to a heavy workload on the available staff who were qualified to provide the necessary input. (H.14) |
| 05000293/FIN-2016004-03 | 2016Q4 | Pilgrim | Failure to Properly Assess and Manage Risk Associated with Shutdown Transformer Protective Relay Testing | Green. The inspectors identified a Green NCV of 10 CFR 50.65(a)(4), Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, for Entergys failure to properly assess and manage the increase in risk due to performing protective relay calibration and functional testing associated with the shutdown transformer (SDT) on seven occasions from December 9, 2005, through August 27, 2014. Specifically, Entergy did not identify that the performance of calibration and functional testing of 6 protective relays associated with the SDT would prevent the 4160V safety buses from being automatically powered by other required sources, and consequently, did not properly assess and manage the increase in risk. Entergys corrective action requires the unit to be in an outage to perform the tests. Entergy entered the issue into the CAP under CR 2017-0856. The inspectors determined that this performance deficiency was more than minor because it is associated with the configuration control attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Additionally, the finding was similar to Example 7e of NRC IMC 0612, Appendix E, Examples of Minor Issues, in that the overall elevated plant risk would have put the plant into a higher licensee-established risk category and would have required additional risk mitigating actions (RMAs). The inspectors evaluated the finding using the Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued October 7, 2016. Because the finding involved a maintenance rule risk assessment, it was screened through IMC 0609, Appendix K, Maintenance Risk Assessment and Risk Management Significance Determination Process, issued May 19, 2005. The finding screened as very low safety significance (Green) using Flowchart 1 of Appendix K because the incremental core damage probability deficit (ICDPD) was determined to be greater than 1E-6 and less than 1E-5, and three or more RMAs were taken. The inspectors concluded this finding had a crosscutting aspect in the area of Human Performance, Avoid Complacency, in that individuals did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk, even while expecting successful outcomes. Specifically, the unavailability of the startup transformer (SUT) and emergency diesel generators (EDGs) during portions of testing was a latent issue that Entergy did not identify, and the associated increase in risk was not assessed and managed. (H.12) |
| 05000293/FIN-2016004-04 | 2016Q4 | Pilgrim | Failure to Correct a Condition Adverse to Quality Associated with Main Steam Isolation Valve | Green. A self-revealing Green NCV of 10 CFR Part 50, Appendix B, Criterion XVI, Corrective Action, was identified in that Entergy did not promptly correct a condition adverse to quality associated with the operability of a MSIV. Specifically, Entergy did not take timely corrective actions to inspect and remove debris from air tubing that supplied air to a valve actuator after the associated MSIV failed a surveillance test on March 29, 2016. This uncorrected condition subsequently led to a repeat failure of the valve on August 16, 2016. Entergy entered these issues into their CAP as CR 2016-2250 and CR 2016-5987 and developed corrective actions to revise associated procedures as needed, replaced the affected MSIV air pack manifold, cleared loose debris from the affected air tubing, and scheduled the replacement of affected air tubing during the next refueling outage. The inspectors determined that this performance deficiency was more than minor because it was associated with the barrier performance attribute of the Barrier Integrity cornerstone and it adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (fuel cladding, reactor coolant system, and containment) protect the public from radionuclide releases caused by accidents or events. Specifically, when MSIV-1C failed to meet its surveillance requirements on March 29, 2016, Entergy did not take corrective actions necessary to adequately identify and resolve the underlying issue of system debris being present in air tubing, which affected the valve actuator and caused a slow closing time for the valve. This inaction led to continued valve inoperability, for a duration greater than that allowed by TS, which presented itself during a subsequent operability test on August 16, 2016. The inspectors screened this finding in accordance with IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power; using Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined that this finding was of very low safety significance (Green) because the finding did not involve an actual open pathway in the physical integrity of reactor containment or involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors determined that this issue had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because Entergy did not use decision-making practices that emphasize prudent choices over those that were simply allowable. Specifically, when the MSIV initially failed its surveillance in March 2016, Entergy did not take a conservative approach in their operability determination and immediate response to the issue. This was demonstrated by the fact that, following the March 2016 valve failure, when a cause evaluation identified the likelihood of debris in air tubing affecting valve operability, individuals rationalized that the degraded condition had been resolved on its own and would not recur. Entergy acted on this assumption, rather than making the conservative determination that the effect of present debris could impact continued operability in an unpredictable manner, as it did during the subsequent failed surveillance test in August 2016. (H.14) |
| 05000293/FIN-2016004-05 | 2016Q4 | Pilgrim | Feedwater Regulating Valve Failure Results in Reactor Scram | Green. A self-revealing Green finding was identified for the inadequate implementation of a work order on the A feedwater regulating valve (FRV) encoder as required by ENWM- 102. Specifically, Entergy did not install a wire assembly on the A FRV encoder as required by the work instructions located in the vendor manual. The wire loosened, resulting in the A FRV failing open and the operators inserting a manual scram. In response to the loose connection, Entergy added a sealant to the connector to ensure all wires remain in place on both FRVs. Entergy entered the issue into the corrective action program (CAP) under condition report (CR) 2016-6635. The inspectors determined that the finding is more than minor because it is associated with the equipment performance attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during a shutdown as well as power operations. Specifically, the performance deficiency affected the reliability and capability of the A FRV which led to a plant scram, tripping of the reactor feed pumps, and closure of the main steam isolation valves (MSIVs). The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, issued October 7, 2016, and IMC 0609, Appendix A, Exhibit 1, Initiating Events Screening Questions, issued June 19, 2012, and determined a detailed risk evaluation was required because the A FRV failure caused a reactor trip and partial loss of feedwater (power conversion system). A Region I senior reactor analyst (SRA) used the Standardized Plant Analysis Risk (SPAR) model for Pilgrim, Version 8.24, and SAPHIRE, Version 8.1.4, to complete the detailed risk evaluation. The estimated increase in core damage frequency (CDF) was calculated to be 4E-7/year, or very low safety significance (Green). For issues resulting in an increase in CDF > 1E-7, IMC 0609 requires an evaluation of large early release frequency (LERF) using the guidance of NUREG-1765, Basis Document for LERF Significance Determination Process, and IMC 0609, Appendix H, Containment Integrity Significance Determination Process, issued May 6, 2004. The performance deficiency associated with the failure of the A FRV and resultant reactor trip would be considered a Type A finding and, as such, the calculated increase in CDF value is used in conjunction with an appropriate LERF factor (multiplier) to determine the estimated increase in LERF associated with the issue. In the absence of early core damage sequences for this event, LERF is not a significance risk contributor and the safety significance of this performance deficiency is defined by the estimated increase in CDF (4E-7/year) or Green. This finding has a cross-cutting aspect of Human Performance, Work Management, in that Entergy did not adequately implement the process of planning, controlling, and executing work activities such that nuclear safety is the overriding priority. Specifically, maintenance staff were provided a work order that did not meet station requirements to ensure the work could be adequately performed. Specific steps of the vendor manual were not used to direct work by staff and led to an installation error. The work planning process also did not implement the engineering recommendation to perform a practice installation on the equipment prior to installing equipment in the field. (H.5) |
| 05000293/FIN-2016004-06 | 2016Q4 | Pilgrim | Licensee-Identified Violation | TS 3.9.B.2 requires that when incoming power is not available from both startup and shutdown transformers, continued operation is permissible, provided both diesel generators and associated emergency buses remain operable, all core and containment cooling systems are operable, and reactor power level is reduced to 25% of design. Contrary to the above, on seven occasions between 2005 and August 27, 2014, for an average of 3.6 hours, Entergy conducted test Procedures 3.M.3-1, A5/A6 Buses 4kV Protective Relay Calibration/Functional Test and Annunciator Verification Critical Maintenance, and 3.M.3-29, Shutdown Transformer and 23kV Relay Calibration and Functional Test, that placed the plant in a condition not allowed by TS 3.9.B.2. Specifically, the testing would have prevented emergency buses A5 and A6 from automatically transferring to their backup power supplies. Entergy entered this condition into their CAP as CR 2016- 2735. A Region I SRA conducted a detailed risk evaluation for this issue using IMC 0609, Appendix A, Significance Determination Process for Findings At-Power, issued June 19, 2012. Using the average time from above, along with operator recovery actions, the SRA calculated the change in core damage probability to be <1E-7, which was considered to be of very low safety significance (Green) |
| 05000293/FIN-2016003-01 | 2016Q3 | Pilgrim | Process Radiation Monitor Subsystems 10 CFR 50.65(a)(2) Not Met | Inspectors identified a Green NCV of 10 CFR 50.65 (a)(2), because Entergy did not adequately demonstrate that the performance of the process radiation monitors (PRMs) was effectively controlled through performance of appropriate preventive maintenance. Specifically, Entergy did not identify and properly account for functional failures of four PRM subsystems in July 2014 and February, April, and July 2015; and did not recognize that the subsystems had exceeded their performance criteria and required a Maintenance Rule (a)(1) evaluation. Entergy entered the issue into the CAP under CR-2016-05564. Entergy performed the Maintenance Rule (a)(1) evaluation, and placed them into (a)(1) where they will be monitored against specific goals. The finding is more than minor because it is associated with the Plant Facilities/Equipment and Instrumentation attribute of the Public Radiation Safety cornerstone and affects the cornerstone objective of ensuring the adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. Specifically, following the failures of the Main Stack Normal Range subsystem in July 2014, the Reactor Building Closed Cooling Water (RBCCW) subsystem in February 2015, the Shared Components subsystem in April 2015, and the Torus Containment High Radiation Monitoring System (CHRMS) subsystem in July 2015, Entergy did not identify the failures as functional failures, and consequently, did not establish goals and monitoring criteria in accordance with 10 CFR 50.65(a)(1). The inspectors determined that the failures demonstrated that the performance of the subsystems was not being effectively controlled through appropriate preventive maintenance, because the incorrect screenings resulted in exceedance of the subsystems performance criteria and placement in (a)(1) status. The inspectors evaluated the significance of this finding using IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process. The finding is of very low safety significance (Green) because the finding was in the Effluent Release Program, but did not result in a failure to implement the Effluent Release Program, and did not result in dose to the public in excess of 10 CFR 50, Appendix I criterion or 10 CFR 20.1301(e) limits. The inspectors determined that the finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, in that the organization did not thoroughly evaluate issues to ensure that resolution addressed causes and extent of conditions commensurate with their safety significance. Specifically, Entergy identified all of the failures of the PRM subsystems, however, Entergy did not thoroughly evaluate the failures as maintenance rule functional failures. |
| 05000293/FIN-2016003-02 | 2016Q3 | Pilgrim | Inadequate Operability Assessment on EDG B | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, in that Entergy did not perform an adequate operability evaluation in accordance with EN-OP-104, Operability Determination Process, Revision 10. Specifically, during an instrumented run of emergency diesel generator (EDG) B, the cabinet door was opened, resulting in a non-seismically qualified configuration of protective relays for EDG B. Inspectors determined that Entergy did not adequately assess the operability of EDG B as required by EN-OP-104, Operability Determination Process. Specifically, Entergy did not evaluate the operability of EDG B when opening a cabinet door containing relays that serve a safety function. Entergy entered this issue into the corrective action program (CAP) as condition report (CR)-2016-5779 and CR-2016-7877. Entergy has issued a standing order to assess operability of equipment tested with cabinet doors open prior to performing work or declare the equipment being tested inoperable. This is a performance deficiency that was within Entergys ability to foresee and correct. This finding is more than minor because it is associated with the protection against external factors attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, relays were no longer in a configuration known to operate as required during a seismic event with the cabinet door open. In accordance with IMC 0609.04, initial Characterization of Findings, and Exhibit 2 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2013, the inspectors determined that this finding is of very low safety significance (Green) because the performance was not a design or qualification deficiency, did not involve an actual loss of safety function, and did not represent actual loss of safety function of a single train for greater than its technical specification (TS) allowed outage time. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Operating Experience, in that the organization did not systematically and effectively collect, evaluate, and implement relevant internal and external operating experience in a timely manner. Specifically, Entergy did not evaluate industry operating experience on control of cabinet doors containing safety-related equipment, which led to operability concerns. |
| 05000293/FIN-2016003-03 | 2016Q3 | Pilgrim | Failure to Properly Implement Agastat Control Relays Preventive Maintenance Procedure in Accordance with TS 5.4.1 | The inspectors identified a Green NCV of Technical Specification (TS) 5.4.1, Procedures, because Entergy did not implement procedures recommended in Regulatory Guide 1.33, Revision 2, Appendix A, February 1978. Entergy did not implement preventive maintenance procedural requirements to periodically replace six high critical, normally energized Agastat EGP relays every 10 years. Entergys immediate corrective actions included replacing all six relays and performing an equipment apparent cause evaluation. Entergy entered this issue into their CAP as CR-2016-04243. The performance deficiency was more than minor because it was associated with the structures, systems, and components (SSCs) and barrier performance attribute of the Barrier Integrity cornerstone and adversely affected the cornerstone objective of providing reasonable assurance that physical design barriers (containment) protect the public from radionuclide releases caused by accidents or events. The failure to replace the relays in accordance with preventative maintenance requirements increased the likelihood of failure for safety systems that relied on these relays for operation. The inspectors determined that this finding is of very low safety significance (Green) in accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2013, because the performance deficiency did not result in an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors determined that this finding had no cross-cutting aspect because the most significant causal factor, the failure to include the relays in the preventative maintenance program database, did not reflect current licensee performance. There was no indication that this specific performance deficiency occurred in the last three years. |
| 05000293/FIN-2016003-04 | 2016Q3 | Pilgrim | Failure to Adequately Evaluate the Effect of Degraded Normally Energized Agastat relays on PCIVs Operability | The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion V, Instructions, Procedures, and Drawings, because Entergy did not perform an immediate operability determination and adequately evaluate the operability of primary containment isolation valves (PCIVs) in accordance with procedure EN-OP-104, Operability Determinations/Functionality Assessments, Revision 10. Entergys immediate corrective actions included electrically deactivating two relays, 16A-K17X11 and 16AK18X11. Subsequently, two PCIVs, CV-5065-91 and CV-5065-92, were closed until all six relays were replaced. Entergy entered this issue into the CAP as CR-2016-04753. The inspectors determined that the performance deficiency was more than minor because it was associated with the Human Performance attribute of the Barrier Integrity cornerstone and adversely affected the objective of providing reasonable assurance that physical design barriers protect the public from postulated radionuclide releases caused by accidents or events. Specifically, Entergy did not perform a timely and adequate operability determination as required by procedure. It took Entergy 74 days and four different operability determinations upon discovery of the degraded relays to finally conclude that PCIVs CV-5065-91 and CV-5065-92 were operable. The inspectors determined that this finding is of very low safety significance (Green) in accordance with IMC 0609.04, initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2013, because it did not result in an actual open pathway in the physical integrity of reactor containment and did not involve an actual reduction in function of hydrogen igniters in the reactor containment. The inspectors determined that this finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence, because Entergy did not initially evaluate the operability of the Agastat relays thoroughly as prescribed in EN-OP-104. Furthermore, Entergy failed to adequately evaluate the effect of the aging Agastat relays pertaining to the PCIVs operability. |
| 05000293/FIN-2016002-02 | 2016Q2 | Pilgrim | Licensee-Identified Violation | 10 CFR 50.65, Requirements for monitoring the effectiveness of maintenance at nuclear power plants, section (a)(4) requires, in part, that before performing maintenance activities, the licensee shall assess and manage the increase in risk that may result from the proposed maintenance activities. Contrary to the above, Entergy did not assess and manage risk on May 17, 2016, when the electric fire pump was removed from service in conjunction with the X-105 instrument air dryer and portions of the D RHR system. Entergy entered this issue into the CAP as CR 2016-3466. The inspectors evaluated the finding using IMC 0609, Attachment 0609.04, Initial Characterization of Findings, issued June 19, 2012. The attachment instructs the inspectors to utilize IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012. The inspectors determined that the finding is of very low safety significance (Green) because the performance deficiency did not affect the design or qualification of a mitigating structure, system, and component, did not represent a loss of system and/or function, and did not represent an actual loss of at least a single train for greater than its TS allowed outage time. |
| 05000293/FIN-2016002-01 | 2016Q2 | Pilgrim | Inadequate Review of Vendor Documents Results in Shear Pin Failure | A Green self-revealing finding was identified for the inadequate design verification of the travelling screens system in accordance with EN-DC-149, Acceptance of Vendor Documents. Specifically, Entergy replaced travelling screens C and D during the May 2015 refueling outage, but did not identify that the installed shear pins did not meet the plant design during engineering reviews of the modification. This caused the shear pins in the C and D traveling screens to prematurely fail during a large seaweed intrusion event on May 5, 2016, which lead to a 50 percent rapid reduction in power. Entergy installed the modified shear pin assembly into the C and D travelling screens per the plant design and restored the screens to service on May 6, 2016. The finding was entered into the corrective action program (CAP) as CR-2016-3202. This finding is more than minor because it is associated with the Initiating Events cornerstone attribute of Design Control and affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. Specifically the failure of the C and D travelling screens shear pins resulted in an unplanned 50 percent reduction in power. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 1 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, the inspectors determined that this finding was of very low safety significance (Green) because the finding did not cause a reactor trip and the loss of mitigation equipment relied upon to transition the plant from the onset of the trip to a stable shutdown condition. The inspectors determined that the finding had a cross-cutting aspect in Human Performance, Avoid Complacency, because Entergy did not recognize and plan for the possibility of mistakes, latent issues, and inherent risk. Specifically, Entergy did not identify that vendor supplied documentation and part numbers did not match Entergys updated documentation. |
| 05000247/FIN-2015004-01 | 2015Q4 | Indian Point | Licensee-Identified Violation | From 2010 to 2014, Indian Point made four shipments of radioactive material that contained category two levels of radioactive material quantity of concern but did not implement a transportation security plan for these shipments, which is contrary to the requirements of 49 CFR 172, Subpart I, Safety and Security Plans. This performance deficiency adversely affected the Public Radiation Safety cornerstone attribute of Program and Process based on inadequate procedures associated with the transportation of radioactive materials. This issue was documented in Entergys CAP as CR-IP2-2015-01985 and HQ-2015-00526. Corrective actions included revision of procedure EN-RW-106 and selection of a vendor to regularly review the federal register for regulatory changes that can impact plant operations. |
| 05000293/FIN-2016010-01 | 2015Q4 | Pilgrim | Failure to perform hourly fire watches | 10 CFR 50.48(a)(1) requires that each holder of an operating license must have a fire protection plan that: (i) describes the overall fire protection program for the facility; (ii) identifies the various positions within the licensee's organization that are responsible for the program; (iii) states the authorities that are delegated to each of these positions to implement those responsibilities; and (iv) outlines the plans for fire protection, fire detection and suppression capability, and limitation of fire damage. Pilgrim Nuclear Power Station (PNPS) Technical Specification 5.4.1.d requires that written procedures shall be established, implemented, and maintained covering Fire Protection Program implementation. ' PNPS implementing procedure 8.8.14, "Fire Protection Technical Requirements," Section 7.5, "Completing Attachment 1 (Hourly Fire Watch)," requires, in part, that fire watch personnel examine the area involved in the posting for evidence of fire or conditions that may lead to a fire. This section further requires that the posting should be visited once every hour such that no fewer than 24 patrols are completed in a 24-hour period at approximately 60-minute intervals. Contrary to the above, on occasions between June 1, 2012, and June 26, 2014, the licensee did not implement a provision of a written procedure covering implementation of the fire protection program as it pertains to fire detection. Specifically, although hourly fire watches were established, fire watch personnel did not examine the areas involved in the hourly fire watch postings for evidence of fire or conditions that may lead to a fire. As a result, for the involved areas, fewer than 24 patrols were completed in 24-hour periods. |
| 05000286/FIN-2015003-01 | 2015Q3 | Indian Point | Blocked Drains in the 480 Volt Switchgear Room | The inspectors identified a Green finding (FIN) because Entergy allowed the Unit 3 480 volt switchgear room floor drains to become blocked such that they could not mitigate an internal flood postulated in Action and Condition Tracking Form 95-14218. Specifically, if both service water (SW) relief valves in the 480 volt switchgear room lifted, their flow rate would be greater than the as-found drain flow rate. This finding does not involve enforcement action because no violation of regulatory requirement was identified. This finding was more than minor because it was associated with the protection against external factors attribute of the Mitigating Systems cornerstone, and adversely affected the cornerstone objective to ensure the capability of systems to respond to initiating events to prevent undesirable consequences. Specifically, the Unit 3 480 volt switchgear room floor drains were not capable of mitigating an internal flood hazard to prevent damage to the 480 volt switchgear, potentially resulting in core damage. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 4, External Events Screening Questions, of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined this finding required a detailed risk assessment. A detailed risk assessment was conducted using the Unit 3 SDP External Event Notebook, which determined that there was a change in core damage frequency of low E-8 per reactor year (an increase in 1 in 100 million reactor years). Therefore, this performance deficiency was of very low safety significance (Green). The inspectors determined the finding does not have a cross-cutting aspect. Although Entergy did not thoroughly evaluate the Unit 2 blocked floor drain issue in 2011 to ensure the resolution addressed extent of condition, Entergy has improved their extent of condition evaluation guidance since 2012. |
| 05000220/FIN-2015009-02 | 2015Q3 | Nine Mile Point | Inadequate Maintenance Rule Monitoring of Unit 1 600 VAC Breaker Super System | The inspectors identified a Green NCV of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance at Nuclear Power Plants, under section (a)(1) and (a)(2) for failing to properly monitor the 600 volt alternating current (VAC) system at Unit 1 in accordance with established maintenance rule reliability criteria to assure that breakers were capable of performing their intended function. Specifically, the inspectors identified four events that were not evaluated against the established (a)(2) reliability criteria. This resulted in a failure to evaluate the 600 VAC system for potential corrective actions in accordance with (a)(1) and did not ensure effective control through preventative maintenance to show the system was capable of performing its intended function in accordance with (a)(2). Exelons immediate corrective actions included evaluations of the failures and planning for a maintenance rule expert panel for consideration of placing the system into (a)(1) where corrective actions could be developed to return the system to (a)(2) monitoring. Exelon also noted that IR 2416790 documented the challenge associated with overcurrent trip device drift and subsequent pump failures. This IR was open at the time of the inspection with actions to determine if a replacement is possible and to present any potential options to Plant Health Committee in October 2015. This performance deficiency is more than minor because it is associated with the equipment performance attribute of the Mitigating Systems cornerstone and affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences. Specifically, the overcurrent trip devices associated with Unit 1 600 VAC General Electric (GE)-AK breakers were unreliable and resulted in the trip of five safety-related pumps between April 2013 and February 2014. Only one of the five functions was evaluated by Exelon. This impacted the ability of these pumps to be able to perform their function to provide cooling to their respective systems. In accordance with IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined this finding was of very low safety significance (Green) because this finding did not represent an actual loss of system safety function, did not represent an actual loss of function of at least a single train for greater than its TS allowed outage time, and did not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with Exelons maintenance rule program for greater than 24 hours. This finding has a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon failed to thoroughly evaluate the failures against the monitoring criteria specified for the Unit 1 600 VAC breaker super system. Specifically, between April 2013 and February 2014, four breaker failures were identified by the inspectors that were not evaluated against the Unit 1 600 VAC breaker super system, which prevented compliance with 10 CFR 50.65 (a)(1) to ensure corrective actions are established to return the system to (a)(2) monitoring. |
| 05000247/FIN-2015003-02 | 2015Q3 | Indian Point | Incorrect Operability Determination Results in Failure to Comply with Technical Specification for Containment Integrity | The inspectors identified a Green NCV for Unit 2 Technical Specification (TS) 3.6.1, Containment, because between August 11 and August 14, 2015, containment out-leakage during accident conditions would have exceeded the containment leakage rate testing program limit specified in TS 5.5.14.c. Specifically, the 24 fan cooler unit (FCU) SW piping developed a hole and Entergys immediate operability determination (IOD) incorrectly concluded that it did not impact operability. Entergy entered this issue into their corrective action program (CAP) as CR-IP2-2015-3550, completed a prompt operability determination (POD) that required compensatory measures, and implemented those compensatory measures on August 14, 2015. This finding is more than minor because it was associated with the configuration control attribute of the Barrier Integrity cornerstone, and adversely affected the cornerstone objective to provide reasonable assurance that physical design barriers, such as containment, protect the public from radionuclide releases caused by accidents or events. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined that the finding was of very low safety significance (Green), because it did not represent an actual open pathway in the physical integrity of reactor containment or heat removal components. For the duration of the violation, SW system pressure remained higher than containment pressure, preventing out-leakage. This finding had a cross-cutting aspect in the area of Human Performance, Conservative Bias, because Entergy did not demonstrate a conservative bias when they assumed the opening in the pipe was too small to impact containment integrity. |
| 05000286/FIN-2015003-04 | 2015Q3 | Indian Point | Licensee-Identified Violation | TS 3.6.6, Containment Spray System and Containment Fan Cooler System, requires two containment spray trains and three containment fan cooler trains to be operable in Modes 1, 2, 3, and 4. TS SR 3.6.6.3 verifies that each containment FCU cooling water flow rate is equal to or greater than 1400 gpm every 92 days. Contrary to TS SR 3.6.6.3, during an essential SW header flow balance test in accordance with 3-PT-R200 on March 3, 2015, three of the five FCUs had coolant water flow less than the required 1400 gpm. Engineering was contacted prior to continuing 3-PT-R200 and directed operations to continue and adjust the FCU throttle valves to obtain FCU outlet SW flow at or greater than 1430 gpm (to account for 30 gpm correction factor for instrument error). On April 9, 2015, during a review of anomalous data identified during 3-PT-R200, Entergy engineering determined that the quarterly surveillance test, 3-PT-Q016, EDG and VC Temperature Valves SWNFCV- 1176 and 1176A and SWN-TCV-1104 and 1105, was not performed with the correct SW system alignment. Entergy identified the cause of the condition was improper implementation of improved TS requirements in 2001. Entergy entered this issue into their CAP as CR-IP3-2015-1063 and CR-IP3-2015-2448. The inspectors evaluated this finding using IMC 0609.04, Initial Characterization of Findings, and IMC 0609, Appendix A, Exhibit 3, Barrier Integrity Screening Questions. The inspectors determined that the finding was of very low safety significance (Green) because the finding did not represent a loss of safety function. |
| 05000247/FIN-2015003-05 | 2015Q3 | Indian Point | Licensee-Identified Violation | Unit 2 TS 5.4.1.a requires that the procedures listed in Attachment A to RG 1.33, Quality Assurance Program Requirements, Revision 2, be established and implemented. Attachment A states that instructions should be prepared, as appropriate, for draining and changing mode of operations for containment cooling systems. NEI 07-07, Objective 1.2.2, requires licensees to evaluate work practices, such as draining of systems that involve licensed material and for which there is a credible mechanism for the material to reach groundwater. Contrary to the above, Entergy did not evaluate work practices involving changing the mode of operation and draining of the containment spray system (a containment cooling system) to assure that the drainage did not reach groundwater; and as a result, during the Unit 2 refueling outage in March 2014, the containment spray system was drained to a floor drain which subsequently overflowed, spread on the floor of a piping room, and leaked through the floor to groundwater. The violation was identified by Entergy in their investigation of groundwater activity identified at the end of the outage during planned sampling of monitoring wells on the site. The issue is a finding as it affected the Public Radiation Safety cornerstone, since Entergys actions resulted in an unintended abnormal effluent release. This finding was assessed using IMC 0609D, Public Radiation Safety, and was determined to be of very low safety significance (Green) because the subsequent groundwater release was a very small fraction of routine liquid radioactive effluent releases, and did not represent any significant dose impact to the public. Entergy documented the issue in their investigation evaluation (CR-IP2-2014-2564) and corrected the issue by revising their draining procedure OAP-038, Operations Mechanical Equipment Operating Guidelines, to assure that contaminated fluids are not discharged outside of the selected drain point. Entergy also provided training to operators on expectations during draining evolutions to assure contaminated liquids are properly controlled. |
| 05000247/FIN-2015003-03 | 2015Q3 | Indian Point | Failure to Conduct Operations to Minimize the Introduction of Residual Radioactivity to the Site | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 20.1406(c) in that Entergy did not conduct operations to minimize the introduction of residual radioactivity into the site. Specifically, Entergy did not identify a new leak of tritium into groundwater based on monitoring well results obtained in February 2015 and did not take action to minimize the introduction of residual radioactivity into the subsurface of the site. Entergy entered this issue into their CAP as CR-IP2-2015-03806 with actions to characterize and evaluate this new leak. The issue is more than minor because it is associated with the program and process attribute of the Public Radiation Safety cornerstone, and adversely affected the cornerstone objective to ensure Entergys ability to prevent inadvertent release and/or loss of control of licensed material to an unrestricted area. In accordance with IMC 0609, Appendix D, Public Radiation Safety Significance Determination Process, the finding was determined to be of very low safety significance (Green) because the issue involved radioactive material control but did not involve: (1) transportation or (2) public exposure in excess of 0.005 rem. The finding had a cross-cutting aspect in the area of Human Performance, Problem Identification and Resolution, in that the resolutions to address the causes for the 2014 tritium leak did not include an extent of condition that recognized the February 2015 tritium spike as a new leak |
| 05000220/FIN-2015009-01 | 2015Q3 | Nine Mile Point | Failure to Identify and Correct a Condition Adverse to Quality Associated with Secondary Containment Leakage | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) 50, Appendix B, Criterion XVI, Corrective Actions, because between 2007 and 2015, Exelon staff did not promptly identify and correct a deficiency associated with Unit 2 reactor building service water pipe penetration W-3177-C. Specifically, on August 20, 2015, during Exelon staffs investigation of an inspector concern associated with the service water pipe penetration into secondary containment, a leakage path into secondary containment was discovered and was not previously identified and evaluated for impact on operability of Unit 2 secondary containment. Exelon generated issue report (IR) 2544831 to document the newly identified condition. The assessment included a review of previously identified leakage paths that were being tracked in accordance with procedure, previously performed secondary containment drawdown leakage testing, and a comparison to the maximum allowable flow rate leakage area. The assessment concluded that based on the gap that was identified at secondary containment penetration W-3177-C, there was a new total of 1.783 square inches of surface area allowing leakage into the Unit 2 secondary containment. Exelon determined this to be acceptable because calculations for secondary containment drawdown testing allows for up to 33.6 square inches of surface area causing in-leakage into secondary containment. Given 1.783 square inches of total identified leakage being less than the allowable 33.6 square inches, there was reasonable assurance that standby gas treatment system will be able to perform its drawdown function and maintain secondary containment vacuum 0.25 inches of vacuum water gauge in accordance with Technical Specification (TS) 3.6.4.1, Secondary Containment. This performance deficiency was more than minor because it impacted the design control attribute of the Barrier Integrity cornerstone objective to provide reasonable assurance that physical design barriers protect the public from radionuclide releases caused by accidents or events. Specifically, Exelons staff failed to identify the degraded penetration seal that impacted the reasonable assurance of Unit 2 secondary containment operability. In accordance with IMC 0609.04, Initial Characterization of Findings, and Exhibit 3 of IMC 0609, Appendix A, The Significance Determination Process for Findings At-Power, issued June 19, 2012, the inspectors determined this finding was of very low safety significance (Green) because the finding only represented a degradation of the radiological barrier function provided for the control room, or auxiliary, spent fuel pool, or standby gas treatment system (i.e., secondary containment). This finding had a cross-cutting aspect in the area of Problem Identification and Resolution, Evaluation, because Exelon staff failed to properly evaluate the condition identified in multiple IRs to determine the extent of condition associated with secondary containment water in-leakage. Specifically, between 2007 and 2015, three IRs were generated and a 2012 structural monitoring review documented the service water penetration water in-leakage and the issue was not appropriately evaluated for the potential for a service water pipe through-wall leak or the potential impact on secondary containment. |
| 05000286/FIN-2015002-03 | 2015Q2 | Indian Point | Incomplete 50.73 Report Associated with Failures of Main Steam Safety Valves | The inspectors identified a Severity Level IV NCV of 10 CFR 50.9(a); in that, Entergy did not provide complete information in a report submitted per 10 CFR 50.73(a)(2)(i)(B). Specifically, a Licensee Event Report (LER) submitted on April 27, 2015, which reported three MSSV test failures (MS-46-2, MS-45-4, MS-47-4) that occurred on February 27, 2015, did not discuss the failure of MSSV 46-3, which also failed its TS as-found lift setting test and was declared inoperable on March 22, 2015. MSSV 46-3 was inoperable for greater than its TS allowed outage time, which is a condition prohibited by TSs, and therefore is required to be reported to the NRC. The inspectors evaluated this performance deficiency in accordance with the Traditional Enforcement process. In accordance with Section 2.2.2.d of the NRC Enforcement Policy, the inspectors determined that the performance deficiency identified with the reporting aspect of the event is a Severity Level IV violation, because it is of more than minor concern, with relatively inappreciable potential safety significance and is related to findings that were determined to be more than minor issues. Specifically, this issue is related to a more than minor corrective action finding, which is documented in Section 1R22 of this report. In accordance with IMC 0612, Appendix B, this traditional enforcement issue is not assigned a cross-cutting aspect. |
| 05000286/FIN-2015002-02 | 2015Q2 | Indian Point | Failure to Maintain the Effectiveness of Emergency Plan Due to an Inadequate Basis for Emergency Action Level Thresholds | The inspectors identified a Green NCV of 10 CFR 50.54(q)(2) for Entergys failure to maintain the effectiveness of an emergency plan that meets the requirements in Appendix E to Part 50 and the planning standards of 50.47(b). Specifically, Entergy did not use accurate facility effluent parameters in its emergency classification and emergency action level (EAL) scheme. Entergy subsequently determined an acceptable facility parameter and corrected the EAL scheme. This finding was determined to be more than minor because it is associated with the Procedure Quality attribute of the Emergency Preparedness cornerstone and adversely affected the cornerstone objective to ensure that Entergy is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The inspectors determined this finding was an emergency classification system finding and therefore evaluated the finding in accordance with IMC 0609, Appendix B, Section 5.4, 10 CFR 50.47(b)(4), Emergency Classification System. The finding was determined to be of very low safety significance (Green) because the issue had a very low likelihood of resulting in an early General Emergency (GE) declaration, but the finding more closely fit the would result in unnecessary classification significance category rather than the would result in unnecessary protective actions for the public significance category. Specifically, the inspectors considered that (1) the inadequate concentration threshold is used as a backup to the effluent threshold which reflects actual plant vent flow, (2) the calculation input discrepancy is small compared to the uncertainty of the setpoint calculation, and (3) although the protective action recommendation (PAR) would be made before offsite dose exceeded the Environmental Protection Agency protective action guideline, an early PAR during an actual release sequence would still serve to provide dose savings to the public. The inspectors determined that the finding has a crosscutting aspect in the area of Human Performance, Challenge the Unknown, because Entergy did not stop when faced with uncertain conditions of the plant vent flowrates and EAL threshold calculation assumptions. |
| 05000286/FIN-2015002-04 | 2015Q2 | Indian Point | Licensee-Identified Violation | On February 27, 2015, MSSV 46-2, 45-4, and 47-4 failed to meet the TS as-found lift setpoint test. Additionally, on March 22, 2015, MSSV 46-3 failed to meet the TS asfound lift setpoint test. TS Limiting Condition for Operation 3.7.1.A, Main Steam Safety Valves, requires if one or more required MSSVs are inoperable, reduce neutron flux trip setpoint to less than or equal to the applicable percent Reactor Thermal Power listed in Table 3.7.1- 1 within 4 hours. If the required action and associated completion time is not met, the reactor shall be placed in Mode 3 in 6-hours and Mode 4 in 12 hours. Contrary to this requirement, MSSV 46-2, 45-4, 47-4, and 46-3 were inoperable for a time period that exceeded the TS allowed outage. This finding was determined to be of very low safety significance (Green) because the finding did not represent a loss of safety function for the MSSV system. This issue was documented in Entergys CAP (CR-IP3-2015-0898 and CR-IP3-2015-2128) and reports were made to the NRC in LER 05000286/2015-002-00 and 2015-002-01. |
| 05000334/FIN-2015008-01 | 2015Q2 | Beaver Valley | Failure to Initiate a Condition Report for an Adverse Condition | A Green self-revealing finding of NOP-LP-2001, Corrective Action Program, was identified after FENOC failed to generate a condition report for a condition adverse to quality. Specifically, FENOC did not initiate a condition report when a lifted lead was identified during preventative maintenance and installation of the Unit 1 main transformer. As a result, corrective actions were not taken and this led to an unplanned downpower from 100 percent to 15 percent reactor power on January 31, 2014. The performance deficiency was more-than-minor because it was associated with the Equipment Performance attribute of the Initiating Events cornerstone, and adversely affected the cornerstone objective of limiting the likelihood of events that upset plant stability and challenge critical safety functions during shutdown as well as power operations. This finding was determined to be of very low safety significance (Green), because it did not cause a reactor trip and the loss of mitigation equipment. This finding has a cross-cutting aspect in the area of Human Performance, Field Presence, because FENOC failed to ensure supervisory and management oversight of work activities, including contractors and supplemental personnel (H.2). |
| 05000286/FIN-2015002-01 | 2015Q2 | Indian Point | Inadequate Corrective Action for Main Steam Safety Valve 46-3 Failure to Lift at Required Setpoint | The inspectors identified a Green NCV of Title 10 of the Code of Federal Regulations (10 CFR) Part 50, Appendix B, Criterion XVI, Corrective Action, for Entergys failure to take corrective actions for a condition adverse to quality involving Unit 3 Main Steam Safety Valve (MSSV) 46-3. Specifically, MSSV 46-3 failed to meet its Technical Specification (TS) required lift setting during a surveillance test on March 22, 2015. This failure was documented in a condition report (CR) but closed for trending purposes. Additionally, Entergy personnel did not correct the failure of MSSV 46-3 to meet its TS required lift setting after it failed its as-found lift setting test on March 1, 2013. The inspectors determined the performance deficiency was more than minor because it is associated with the Equipment Performance attribute of the Mitigating Systems cornerstone and adversely affected the cornerstone objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, Entergy did not take corrective actions following the March 22, 2015, failure of MSSV 46-3, and previous corrective actions in 2013 were not effective in ensuring it would remain capable of lifting at its TS required setpoint. The inspectors determined that this finding is of very low safety significance (Green) because the finding does not represent an actual loss of function of one or more non-TS trains of equipment designated as high safety-significant in accordance with Entergys maintenance rule program for greater than 24 hours. Specifically, of the 20 valves tested in 2015, 16 passed the as-found lift test and there was no loss of safety function. The inspectors determined that this finding had a Problem Identification and Resolution cross-cutting aspect related to Evaluation, because Entergy did not thoroughly evaluate issues to ensure that resolutions address causes and extent of conditions commensurate with their safety significance. Specifically, the CR documenting the MSSV 46-3 failure was closed for trending purposes and as a result, a thorough evaluation of the cause was not completed. |
| 05000247/FIN-2015001-02 | 2015Q1 | Indian Point | Untimely Corrective Actions for Degraded Fire Protection Piping Results in Piping Break | The inspectors identified a self-revealing NCV of license condition 2.K. because Entergy did not take adequate corrective actions for degraded fire protection piping in the Unit 1 turbine building. This issue contributed to excessive leakage and failure of a 10-inch high-pressure fire protection spool piece. Depressurization and isolation of this leak resulted in loss of high-pressure fire water to Unit 2 until compensatory measures could be established after about two hours. Entergy entered this issue into their CAP as CR-IP2-2014-6668, repaired the piping section, and is prioritizing repairs to other sections of degraded piping. This finding is greater than minor because it adversely affected the Mitigating Systems cornerstone objective to ensure the availability and reliability of systems (fire protection system) that provide protection against external events (fire) when all the fire protection pumps were secured to isolate the failed piping. This finding was evaluated using IMC 0609, Appendix F, Fire Protection Significance Determination Process, question 1.4.7, Fire Water Supply. It was found to be of very low safety significance because at least 50 percent of the fire water capacity (5500 gpm) remained available when the leak occurred. The inspectors determined that this finding had a cross-cutting aspect in Problem Identification and Resolution, Resolution, because Entergy did not take effective corrective actions to address issues in a timely manner commensurate with their safety significance, resulting in the piping break (P.3). |
| 05000247/FIN-2015001-01 | 2015Q1 | Indian Point | Failure to Control Transient Combustibles in Accordance with the Approved Fire Protection Program | The inspectors identified an NCV of the license condition 2.K. when Entergy failed to properly control transient combustibles within the Unit 2 control room envelope in accordance with the approved fire protection program (FPP). The inspectors identified transient combustible material in excess of the specified limits that were unattended and without a transient combustible evaluation (TCE). The inspectors notified Entergy personnel of the deficiency, the transient combustibles were promptly removed, and the issue was entered into the corrective action program (CAP) as condition report (CR)-IP2-2015-1058. The inspectors determined that the failure to properly control transient combustible material in accordance with the approved FPP was a performance deficiency. This finding was determined to be more than minor because it is associated with the protection against external factors attribute of the Initiating Events cornerstone and adversely affected the cornerstone objective to limit the likelihood of events that upset plant stability and challenge critical safety functions during power operations. In accordance with IMC 0609.04, Phase 1 Initial Characterization of Findings, the inspectors determined that the finding affected the administrative controls for transient combustible materials. The inspectors conducted a Phase 1 SDP screening using IMC 0609, Appendix F, Fire Protection Significance Determination Process, and assigned the finding to the Fire Prevention and Administrative Controls category; in that, it affected Entergys combustible materials control. The finding was determined to be Green, or very low safety significance, after IMC 0609, Appendix F. question 1.3.1, Is the reactor able to reach and maintain safe shutdown (hot or cold) condition, was answered yes. The inspectors assumed that any fire in the area associated with the combustibles observed would be promptly extinguished using readily available extinguishing equipment and that no safety-related equipment would be disabled. The inspectors determined that this finding had a Human Performance, Procedure Adherence, cross-cutting aspect because Entergy failed to properly control transient combustible material in accordance with the approved FPP when the allowed limits were exceeded without an evaluation (H.8). |
| 05000286/FIN-2015001-03 | 2015Q1 | Indian Point | Licensee-Identified Violation | TS LCO 3.4.13, RCS Operation Leakage, states, in part, that RCS operational leakage shall be limited to no pressure boundary leakage. Contrary to the above, on March 14, 2013, during a scheduled RFO boric acid program walk down inspection, Entergy identified a through-wall defect and therefore a RCPB leak on a fillet weld which attaches the E-11 in-core guide tube to the seal table path. Corrective actions included a VT-2 visual examination of the remaining seal table penetrations to verify that no additional through wall leaks were present. Additionally, the leaking guide tube was removed from service by cutting the tube below the leaking area and installing a welded plug to form a new RCPB. No performance deficiency was identified because it was not reasonable for Entergy to foresee and prevent the pressure boundary leak. Since this violation has no performance deficiency, traditional enforcement applies. The inspectors evaluated the significance of the issue using traditional enforcement and determined it was a SL IV NCV of TS 3.4.13 in accordance with the NRC Enforcement Policy, Section 6.1.d. This issue was entered into Entergys CAP as CR-IP3-2013-01556 and a report was made to the NRC in LER 05000286/2013-004-00. |
| 05000333/FIN-2014005-01 | 2014Q4 | FitzPatrick | Untimely 10 CFR 50.72 Notification of a Secondary Containment System Functional Failure | Severity Level IV. The inspectors identified an SL IV NCV of Title10 of the Code of Federal Regulations (10 CFR) 50.72, Immediate Notification Requirements for Operating Nuclear Power Reactors, because unplanned inoperability of the secondary containment system was not reported to the NRC within eight hours of the occurrence, as required by 10 CFR 50.72(b)(3)(v), Event or Condition that Could Have Prevented Fulfillment of a Safety Function. Specifically, while restoring the normal reactor building ventilation (RBV) syste to service following maintenance, reactor building-to-ambient differential pressure droppe below the Technical Specification (TS) required minimum value of 0.25 inches of vacuum water gauge and therefore caused the secondary containment system to be inoperable. However, FitzPatrick staff did not promptly recognize this as a condition reportable under 10 CFR 50.72. As corrective action, FitzPatrick staff reported the condition to the NRC in accordance with 10 CFR 50.72(b)(3)(v) and entered it into the corrective action progra (CAP) as condition report (CR)-JAF-2014-06498. The inspectors determined that the failure to inform the NRC of the secondary containment system inoperability within eight hours in accordance with 10 CFR 50.72(b)(3)(v) was a performance deficiency that was reasonably within Entergys ability to foresee and correct. The inspectors evaluated this performance deficiency in accordance with the traditional enforcement process because the issue impacted the regulatory process, in that a safety system functional failure was not reported to the NRC within the required timeframe, thereby delaying the NRCs opportunity to review the matter. Using Example 6.9.d.9 from the NRC Enforcement Policy, the inspectors determined that the violation was a SL IV (more than minor concern that resulted in no or relatively inappreciable potential safety or security consequence) violation, because Entergy personnel failed to make a report required by 10 CFR 50.72 when information that the report was required had been reasonably within their ability to have identified. In accordance with IMC 0612, Power Reactor Inspection Reports, traditional enforcement issues are not assigned cross-cutting aspects. |